Multi-pass, vapor moderated and cooled nuclear reactor and method of operating to variably moderate and control same



3,247,073 VAPOR MODERATED AND COOLED NUCLEAR REACTOR AND CONTROL SAME 5Sheets-Sheet l D. C. SCHLUDERBERG INVENTOR. 32 Donald C. SchluderbergATTORNEY -D1STANC E FROM CORE CENTER 3,247,073 MULTI-PASS, VAPORMODERATED AND COOLED NUCLEAR REACTOR AND METHOD OF OPERATING TO VARIABLYMOD D. C. SCHLUDERBERG ERATE AND CONTROL SAME 5 Sheets-Sheet 2 6 2 6 9 l1 I 3 9 2 l W l M .1 d r. n. m A m INVENTOR. Donald C. SchluderbergATTORNEY Aprll 1966 D. c. SCHLUDERBERG 3, 7, 3

MULTI-PASS, VAPOR MODERATED AND COOLED NUCLEAR REACTOR AND METHOD OFOPERATING T0 VARIABLY MODERATE AND CONTROL SAME Filed May 23, 1962 5Sheets-Sheet 3 FIG.6

JNVENTOR. Donald C. Schluderberg ATTORNEY April 19, 1966 o. c.SCHLUDERBERG 3, 4 ,0 3

MULTI-PASS, VAPOR MODERATED AND COOLED NUCLEAR REACTOR AND METHOD OFOPERATING TO VARIABLY MODERATE AND CONTROL SAME Flled May 23, 1962 5Sheets-Sheet 4 a J I' H: i I :l 160 T V 182 m v ,W, 70 I ma s! 68 160 II :5 -L A-45QEMMMU J INVENTOR.

Donald C. Schluderberg AT TORNEY April 19, 1966 MULTI-PASS, VAPORMODERATED AND COOLED NUCLEAR REACTOR AND METHOD OF OPERATING TO VARIABLYMODERATE AND CONTROL SAME Filed May 25, 1962 D. C. SCHLUDERBERG 5Sheets-Sheet 5 v r 1' e P LB/FT3 v=o.o5, =20 5000 AP Q /l '1 1 -4000 41/ I Y a 1i 8 I .7 Ah A I 6. 7 u I J q a (I L J F I 6 D /Ah 1 w 79 83000 $5 i ,9 g Ar, I ?,A a N & V

ENTHALPY, h BTU/LB FlssloN RANGE OF RESONANCE REACTIONS THERMAL ENERGY ER Y g RANGE RANGE z 1 0 m l l g 1 n: F m 2 "64 m 65 5 J J m :1

NEUTRON ENERGY INVENTOR. Donald C. Schluderberg BY ATTORNEY UnitedStates Patent MULTl-PASS, VAI'0R MODERATED AND QQGLED NUQLEAR REACTORAND IWETHQD 0F (WER- ATJING T0 VARHABLY MDDERATE AND CON- TRGL SAMEDonald C. Schluderberg, Lynchburg, Va, assigner to The Babcock & WilcoxCompany, New York, N.Y., a corporation of N ew .Fersey Filed May 23,1%2, Ser. No. 197,060 12 Claims. (Q1. 176-42) The present inventionrelates in general to a nuclear reactor and, more particularly, to amethod of operating such a reactor wherein a hydrogen-bearing vapor isused to variably moderate, control and cool the chain reaction.

In every nuclear reactor a quantity of fissionable material must bearranged as a core with sufiicient mass and proper configuration toestablish and sustain a fissiontype chain reaction. Further, innon-breeder types of reactors it is necessary to provide fissionablematerial in the core in excess of that necessary merely to establsh thechain reaction in order to sustain the reaction for a practical lengthof time. Such excess fuel is necessary to compensate for the fissionablematerial which is consumed throughout the life of the reactor and toovercome the effect of the accumulation of neutron absorbing poisonmaterials generated by the fission process. As a result of the inclusionof this excess fuel within the core, more neutrons are produced at anygiven instant than are necessary to maintain the self-sustainingfissiontype chain reaction. Accordingly, it is necessary to control thisexcess nurnber of neutrons either by capturing them non-productively,allowing them to escape from the reactor, or by capturing them infertile nuclear material so as to generate or breed additional nuclearfuel.

In reactors of the prior art, nonproductive absorption of excessneutrons has been achieved by introducing into the reactor core neutronpoison material, i.e., a material which absorbs neutrons. Neutron poisonmaterials, such as hafnium or boron, have been introduced into reactorsin the form of movable control rods or, less frequently, introducted inthe form of poison material either built permanently into the structuralmaterial of the reactor core or in a soluble form mixed with the reactorcoolant and circulated through the core. The use of such control rodswithin a reactor necessitates utilization of intricate mechanicalequipment to control and regulate their movement into and out of thereactor core. These control rods increase the cost and complexity of thereactor due to this intricate nature. Furthermore, the utilization ofsuch movable control rods Within a reactor core creates the undesirableeifect of flux peaking in the core which is due, in part, to the factthat the poison material of the control rod is movably positioned withinthe core in the path of the neutrons generated by the chain reactionthereby resulting in peaks of temperature and heat flux within the core.Inasmuch as the maximum surface temperature of fuel elements in the coremust be calculated with due consideration directed to this peaking,reactors are necessarily designed with lower average surface temperatureand/ or heat flux throughout the core than would otherwise be possible,since it is necessary to maintain the maximum possible fuel clad surfacetemperature within safe use limits.

As is well known in the art, the process of fissioning nuclear fuelatoms to produce a chain reaction generates neutrons having a wide range.of high velocities. It is also .known that, generally' the chainreaction is best promoted by slower or thermal neutrons rather than bythe fast neutrons generated by the fissioning of an atom. This resultsfrom the fact that thermal neutrons have a greater probability offissioning new atoms than do the fast neutrons which tend to completelyescape from the reactor core. It is thus necessary to moderate or slowthe neutrons generated by the chain reaction so as to increase thenumber of thermal neutrons available to further promote and sustain thechain reaction. In reactors of the prior art, this moderation isaccomplished by the use of such materials as graphite, beryllium oxide,heavy or ordinary water distributed throughout the core.

In these prior art reactors it has been necessary to use separate anddistinct systems or apparatus to moderate the neutrons created by thechain reaction and to control the chain reaction. This has beennecessary since it has not previously been possible to both moderate andsatisfactorily control the reactor utilizing the same apparatus. Controlrods have been especially necessary in order to insure a safe, reliablemethod of starting, operating and shutting down the reactor. It isrecognized that if a safe dependable method of supplying such controlcould be accomplished without the use of control rods, the flux peakingdiscussed above could be minimized, thereby making possible theutilization of a higher average core temperature with consequentincrease in the final temperature of the cooling medium or an increasein power output for the same final temperature, both resulting inimproved cycle efiiciencies.

Copending application Serial No. 158,022 filed December 8, 1961, inwhich applicant is a coinventor discloses a means both to variablycontrol and moderate a nuclear reactor to produce a self-sustainingchain reaction by introducing a hydrogen-bearing moderating vapor intothe core and regulating the reaction by varying the concentration of thehydrogen constituent in the core. This is accomplished, as disclosed,either by varying the concentration of the hydrogen within themoderating vapor or by varying the density of the moderating vaporwithin the reactor core with a resultant variation in the concentrationof hydrogen therein. In the method disclosed in this copendingapplication it is necessary, in order to achieve operating stability, toutilize separate circuits for the cooling fluid and for the moderatingvapor. At that time, it was not known how to achieve the negativetemperature coefficient of reactivity necessary to attain adequatereactor and system stability when utilizing a single fluid for both thecoolant and the moderating vapor. It has been found in the presentinvention that such utilization of a single fluid both for cooling andfor moderating a reactor is possible if thermal feedback is introducedinto the cooling and moderating vapor. The thermal feedback provides thereactor with a negative temperature coeificient of reactivity so thatupon slight disturbances in reactivity the reactor is self-compensating.Additionally, with the utilization of such thermal feedback it is foundthat the reactor can be made self-regulating, so that upon changes insystem output calling for a change in reactivity, the reactor willautomatically seek the new power output level.

Accordingly, the present invention discloses a method for variablycontrolling, moderating and cooling a nuclear reactor having amultiplicity of fissionable material-bearing fuel elements arranged as acore to undergo a selfsustaining fission-type chain reaction comprisingthe steps of introducing a hydrogen-bearing moderating and cooling vaporinto said core and regulating the fission-type chain reaction by varyingthe concentration of hydrogen in said core.

Additionally, the concentration of hydrogen in the core may be varied byeither varying the concentration of hydrogen in the vapor and/or byvarying the density of the vapor within the core.

Furthermore, the method comprises the step of intro ducing thehydrogen-bearing moderating and cooling vapor into a first group of fuelelements within the core whereby the vapor is heated to a moderatingdensity, then passing the vapor at the moderating density about all ofthe fuel elements in the core to moderate the chain reaction, and thenpassing the vapor through a second group of fuel elements to be furtherheated.

In addition, the present invention relates to a nuclear reactorcomprising a pressure vessel containing a plurality of fissionablematerial-bearing fuel elements arranged therein to undergo aself-sustaining fission-type chain reaction, the fuel elementscomprising a first group and a second group whereby the first group isserially connected in a fluid flow sense between an inlet to thepressure vessel and a moderating space arranged around both the firstand the second groups of fuel elements and the second group of fuelelements is serially connected between the moderating space and thepressure vessel outlet, the first group of fuel elements being suspendedwithin shroud tubes arranged to separate the fuel elements from themoderating space and supported from the upper portion of the pressurevessel and terminating in the lower portion of the moderating space, andthe second group of fuel elements being suspended in a second group ofshroud tubes which are supported from the lower portion of the pressurevessel and which terminate in the upper portion of the moderating space.

The various features of novelty which characterize the invention arepointed out with particularity in the claims annexed to and forming apart of this specification. For a better understanding of the invention,its operating advantages and specific objects attained by its use,reference should be had to the accompanying drawings and descriptivematter in which the preferred embodiments of the invention have beenillustrated and described.

Of the drawings:

FIG. 1 is a schematic illustration of the reactor system of the presentinvention;

FIG. 2 is a vertical section of a specific arrangement of the reactor ofthe present invention;

FIG. 3 is a portion of the cross-sectional view of the specificarrangement of the reactor in the present invention taken along lines3-3 of FIG. 2;

FIG. 4 is an enlarged plan view of a single fuel element bundle takenfrom the cross-sectional view of the reactor as shown in FIG. 3;

FIG. 5 is an enlarged view of a portion of the core of the reactorillustrated in FIG. 2;

FIG. 6 is a cross-sectional view of a portion of a fuel elementillustrating the first, second and third passes of the reactor and takenalong line 66 of FIG. 5;

FIG. 7 is an enlarged cross-sectional view of a portion of the fuelelement within a third pass shroud tube;

FIG. 8 is an enlarged cross-sectional view of a portion of a fuelelement within a first pass shroud tube;

FIG. 9 is an enlarged cross-sectional view of a portion of a fuelelement in an alternate type third pass shroud tube;

FIG. 10 is a graphical representation of neutron abundance versusneutron energy;

FIG. 11 is a graphical representation of steam density corresponding tovarious pressures and enthalpies;

FIG. 12 is a graphical representation or relative power density acrossthe reactor core cross section; and

FIG. 13 is a schematic illustration of an alternate reactor systemutilizing the present invention.

The specific reactor arrangement illustrated in FIGS. 2 to 9 forms theinvention claimed in the copending joint application of J. W. Ryon andD. C. Schluderberg, Serial No. 197,082, filed May 23, 1962.

A schematic illustration of a reactor system of the present invention ispresented in FIGURE 1 wherein a reactor 28 is provided with a suitablepressure retaining vessel 22, the interior of which is divided intothree separate and distinct passes. These passes include a first pass 24through a portion of the reactor, a second or moderating pass 26, and athird pass 28 through the remaining portion of the reactor as will bemore fully described. Fissionable material-bearing'fuel elements, of atype well known in the art, are arranged within the reactor pressurevessel 22 to form a core with a critical mass to undergo aself-sustaining fission-type chain reaction when moderated, as will bedescribed. A portion of the fuel elements are disposed within the coreof the reactor in the first pass 24, and the remaining portion aredisposed in the third pass 28 of the core. These passes are arranged sothat a fluid which is both moderator and coolant, supplied to thereactor through an inlet line 30, flows through the first pass 24,cooling the fuel elements therein by absorbing heat therefrom. Afterbeing heated in the first pass the coolant fluid then flows through thesecond, or moderating pass or space 26 which encompasses the exterior ofthe fuel elements in both the first and third passes. In passing throughthe moderating space the moderating and coolant fluid moderates theneutrons therepresent continuing the chain reaction as will be furtherdescribed hereinbelow. After passing through the moderating space 26 themoderating and cooling fluid passes through the third pass 28, coolingthe fuel elements disposed therein, being heated by direct heat transferto a predetermined outlet temperature. As illustrated schematically inFIGURE 1, the first and third passes 24 and 28, respectively, within thereactor pressure vessel may be comprised of a plurality of tubesdisposed within the reactor with the first pass tubes being connectedwith the inlet line 30 and opening at their opposite end to themoderating space 26, while the third pass tubes connect directly withthe outlet line 32 and open at their opposite ends to the opposite endof the moderating space 26 from that in communication with the open endof the first pass 24. In this way the coolantmoderator enters thereactor 20 through inlet line 30', passes downwardly through the firstpass 24 and enters the lower end of the moderating space 26. Thecoolantmoderator then passes upwardly through the moderating space 26around both the first and third pass tubes and enters the upper end ofthe third pass 28, flowing downwardly therethrough and out of thereactor by outlet line 32.

The heated moderator-coolant fluid then passes through heat exchangers34 and 36 giving up heat to a secondary heat transfer fluid in anindirect heat transfer relation, the latter fluid being transported to apoint of use, not shown. These heat exchangers may operate in series orin parallel for imparting heat to the secondary fluid or one may serveas a superheater 34 and the otheras a reheater 36 for the secondary heattransfer fluid system in a manner well known in the art. Upon leavingthese heat exchangers 34 and 36 the moderator-coolant fluid enters theinlet of pump 4%) which discharges the fluid through line 42 to heatexchanger 46 which acts as a boiler or steam generator for the secondaryfluid. Upon leaving heat exchanger 46 the coolant-moderator fluidreturns to the reactor via inlet line 30.

An orifice 48 is positioned in the pump outlet line 42 to produce aslight pressure drop in the moderator-coolant fluid flowingtherethrough. On either side of this orifice 48, a connection is made bylines 50 and 52 to a surge chamber 54, whose purpose will be furtherdescribed hereinbelow.

A moderator-coolant fluid supply line 56 is provided in the reactorsystem. It is equipped with a metering device 58 and opens into the line38 entering pump 40. An exhaust line at having a metering device 62 anda relief valve 63 is also provided and is in communication with theoutlet line 32. The metering devices 58 and 62 may be of any type Wellknown in the art and may include metering valves or positivedisplacement pumps, the primary requirement of each being that they becapable of transferring accurate, predetermined amounts ofcoolantmoderator fluid into or out of the reactor circuit, as necessary,to permit a closely controlled variation of the concentrationor'inventory of the moderator-coolant fluid within the reactor circuit.Relief valve 63 may be either automatic or selectively operable asconditions dictate to vent fluid from the reactor circuit to a lowerpressure receiver (not shown), thus enabling the pressure within thereactor circuit to be quickly reduced and thereby decreasing theconcentration or inventory of the moderator-coolant fluid Within thereactor in order to terminate the chain reaction.

The moderator-coolant fluid may be any hydrogenbearing vapor including,but not limited to, vaporous hydrocarbons, pure hydrogen gas, steam ormixtures of any of these with inert gases. For reasons later discussedthe preferred moderator-coolant fluid and the one which will bediscussed to a greater extent hereinbelow is steam provided atsupercritical pressures.

It is known that in a nuclear reactor, when the fuel material undergoesfissioning, neutrons having a wide range of energies are generated. Asshown in FIGURE 10, there is a comparatively wide variation in therelative abundance in neutrons produced with respect to the energiesinvolved. Neutrons generated by the fissioning of an atom have a highenergy level and as they move outwardly from the atom being fissionedthey are moderated or slowed by the various materials within thereactor, including both the structural as Well as the moderatingmaterial. As these neutrons are slowed they pass through a resonancereaction energy range, wherein fertile material may capture the neutronsand thus create additional fissionable material out of fertile material,as is Well known in the art. At lower neutron energies, i.e. below theresonance reaction range, is the thermal energy range wherein mostfissionable material reacts with the neutrons therepresent to undergo afission reaction. It has been found that the relative abundance ofneutrons within the range of resonance reactions and within the thermalenergy range may be altered by varying the amount of moderation to whichneutrons are subjected. Thus, if the neutrons are subjected to acomparatively great amount of moderation, the relative abundance ofneutrons within the various energy ranges will approximate that shown byline trace 65 in FIGURE 10. As may be seen, the neutron abundance withinthe thermal energy range is relatively high, while that within theresonance reaction range is relatively low. Conversely, should theamount of moderation be relatively low, the abundance of neutrons withinthe various energy ranges will approximate that indicated by trace line64. Accordingly, during initial operation of a reactor, while the amountof fuel material in the core is relatively high and the amount ofneutron poisons contained therein is relatively low, the chain reactionmay be sustained with less neutron moderation since fewer neutrons arenecessary in the th e 1;mal energy range. After prolonged periods ofoperation, however, the amount of fuel Within the core will havesubstantially decreased with a corresponding rise in fission productneutron poison content within the core. In this situation more neutronmoderation is required since more neutrons are necessary within thethermal energy range to sustain the fission-type chain reaction. As aresult, the relative number of neutrons available within the range ofresonance reactions to react with the fertile material within the core,is reduced.

Thus it may be seen that the efliciency of neutron utilization within areactor may be enhanced throughout the life of the core by the combineduse of a fertile nuclear material with the fissionable nuclear materialin the reactor core in combination with the control of neutronmoderation. This economy is made more apparent when it is realized thatif such fertile material were not utilized in a reactor core, the amountof neutrons made available for the fissioning of the reactor fuel, i.e.in the thermal energy range, would be regulated only by changing thenumber of neutrons absorbed nonproductively by control poisons withinthe core, or by varying the number of neutrons permitted to escape fromthe periphery of the core. In either case utilization of the neutronsgenerated by the chain reaction would be less than optimal and theover-all etficiency less than the maximum attainable when using theteachings of this invention.

It has been found that a reactor of the present invention may beoperated and its output controlled by the utilization of ahydrogen-bearing vapor. While gases, for example pure hydrogen, could beused as the moderatorcoolant of the present invention, the pressurerequired to attain the requisite density within the reactor core toprovide sufficient moderation for sustaining a chain reaction wouldexceed practicable structural limitations. Furthermore, the use of purehydrogen would not provide the desired heat transfer characteristics topermit the construction of an eflicient power producing reactor. It hasbeen found, however, that high. pressure, high temperature steam meetsthe prerequisites for a hydrogen-bearing moderator-coolant. team havinga temperature in the order of 700 F. and a pressure of approximately3400 p.s.i., will provide a suflicient concentration of hydrogen withinthe reactor core to satisfactorily sustain a chain reaction, Whileproviding the desired heat transfer and transport characteristics. Byvarying the density of the steam within the reactor core, by changingits temperature, its pressure, or by diluting the steam with anonmoderating Vapor,

or by any combination of these, the moderation of the chain reaction,and thus the reactivity of a reactor, may be controllably varied.Accordingly, the effective variable moderation of the reactor asdiscussed above with respect to FlGURE 10 is possible using this methodof moderation.

As noted above, in reactors of the prior art control rods have beennecessary in order to provide safety shutdown control of a reactor. Thissafety shutdown control for a nuclear reactor of the present inventionis provided by the relief valve 63 discussed above. Inasmuch as areactor will not operate unless there is suflicient moderation to permitcontinuance of the chain reaction, if the concentration of themoderator-coolant within the reactor is suddenly decreased, the reactorwill quickly become subcritical and shut itself down. Such a decrease inconcentration of moderator-coolant within the reactor is achieved by theopening of the relief valve 63 which vents the reactor fluid to a lowerpressure receiver (not hown). This mode of operation is possible with areactor of the present invention due to the fact that themoderator-coolant fluid is in a vaporous state and so is characterizedby significant changes in density for comparatively small changes inpressure. Where the moderator-coolant is a liquid, as for examplepressurized water, the change in pressure achieved by opening a reliefvalve would not produce Suthcient change in the moderator density toprovide effective control since water in the liquid state is nearlyincome pressible. Neither is such control possible in boiling waterreactors, since the change in pressure efiected by venting would notdecrease the density of the moderating fluid but would only flash tosteam a portion of the boiling liquid therepresent, leaving theremaining liquid in the core to moderate the chain reaction.

In initiating operation of a reactor of the present invention thereactor system is first brought to an equilibrium operating temperatureby circulating a small amount of steam therethrough while continuouslysupplying heat to it from an external heat source (not shown). When thereactor system has reached initial operating temperature, additionalsteam is introduced into the circuit through the inlet metering device58 by way of inlet line 56 thereby increasing the density of the vaporwithin the reactor core, with the neutron moderation increasing to thepoint where reactor criticality is reached and a self-sustainingfissiontype chain reaction is initiated. The reactor power is thenraised slowly to a point where the external heat source is no longerrequired. These and subsequent power increases are achieved byintroducing additional vapor into the system and/or by increasing thehydrogen atom concentration of the vapor.

The heat exchangers 34, 36 and 46 in the reactor system in the meantimecommence their functioning, the primary or moderator-coolant fluidgiving up heat produced within the reactor to a secondary fluid for theconversion to useful work, as for example power, or for some otheruseful end result.

Referring to FIG. 11 it may be seen that for a reactor, which is asubstantially constant volume system, requiring a density p ofmoderating steam of approximately pounds per cubic foot and operatingwithin a temperature range between its saturation temperature and 760F., the moderator-coolant pressure falls within a range of pressures,AP, which extends from about 2800 p.s.i.a. to 3700 p.s.i.a. provided theenthalpies are maintained within a corresponding range of Ahs from about1050 B.t.u./lb. to 1125 B.t.u./lb. The relationships for theseconditions are indicated generally by reference letter W. As the reactorages, the amount of moderation necessary to sustain the chain reactionincreases and the mass of the moderating vapor in the loop would beincreased so that the density of the vapor at the end of core life wouldbe approximately 20 pounds per cubic foot. At that time themoderator-coolant vapor, while being maintained Within the sametemperature range as before, would be operated over a range of pressuresAP, of 3200 p.s.i.a. to 4450 p.s.i.a. with a corresponding range ofenthalpies Ah of 900 Btu/lb. to 960 B.t.u./lb.; as indicated byreference letter Y. While these values have been given in the way ofexamples, it will be appreciated that the density of steam or othermoderator-coolant fluid will be dictated by the particular design of thereactor in question. However, it should be noted that the permissibleupper temperature limit of the moderator-coolant vapor in the moderatorspace 26 will be approximately 760 F. depending upon the amount of heatthat will be transferred to the moderator-coolant vapor by the fuelelements in the third pass and by the final outlet temperature desired.Such final outlet temperature would, of course, be determined by thetemperature limitations of the structural material of the various systemcomponents. The lower temperature limit has been indicated as saturationdue to the requirement that vapor only be contained within the reactorcircuit. However, the lower limit of the moderator-coolant vapor may bedictated by the desire to maintain the moderator-coolant within thereactor in the supercritical pressure range, since when operating atapproximately 3500 p.s.i.a. steam in the temperature range of from 690F. to 1050 F. can absorb 700 B.t.u./lb. available for transference inthe heat exchanger as contrasted to the 400 B.t.u./lb. heat transfercapacity of steam when operating at only 2800 p.s.i.a. As a result ofthis increased heat transfer capacity per pound of steam, each of thesystem 8 components can be made smaller in size for a correspondmg poweroutput.

When it is desired to operate with the moderatorcoolant vapor in thesupercritical range throughout the life of the reactor, it may becomenecessary, in order to achieve satisfactory service life of the nuclearfuel elements, to mix the moderator-coolant vapor with a compatiblevapor having little or no moderating effect upon the chain reaction.Such a diluting vapor or gas could be any substance which is compatiblewith the moderatorcoolant vapor and produces the desired thermodynamiceffects, as for example, heavy water, D 0, which has a moderating effectconsiderably less than that of ordinary water. By dilution of themoderator-coolant vapor satisfactory service life of the nuclear fuelelements can be achieved by varying the system pressure between 3600p.s.i.a. and 3900 p.s.i.a., thus staying within the supercritical range.However, should the density of the mod erator-coolant be variedthroughout the life of the core by varying the pressure and temperatureonly and not by dilution as set forth above, it would be necessary touse a wider range of pressures, for instance between 2800 p.s.i.a. and3900 p.s.i.a., in order to attain satisfactory life from the fuelelements, thereby reducing the heat transport capacity of themoderator-coolant steam as set forth above. Additional benefits arederived from maintaining the moderator-coolant steam in thesupercritical pressure range in that the heat transfer characterlsticsof steam in the temperature range encountered within the first pass fuelelement is almost the same as for water. Accordingly, the structuralmaterial in the first pass fuel elements can be maintained .at atemperature low enough to use material having a low neutron absorptioncross section, thereby reducing the amount of neutron absorbingstainless steel required in the core. In a nuclear reactor it isdesirable to provide a negative temperature coeflicient of reactivitydependent upon the temperature of the moderator-coolant so the reactorwill tend to be self-regulating, adjusting the reactivity level to thepower output demand of the system and minimizing reactivity surgesgenerated within the reactor itself. Keeping in mind the fact that thereactor circuit is a substantially contant volume system, as the outputdemand of the system increases, reducing the temperature of the steamreturning to the reactor by inlet line 30, the moderator-coolant vaporwill have a higher density entering the reactor resulting in greaterneutron moderation and an increase in the power output of the reactor.Conversely, should the output demand of the system decrease, themoderator-coolant steam temperature entering the reactor will be raised,since the heat exchangers are rejecting some available heat, therebydecreasing its density and its moderating effect and reducing the poweroutput in the reactor to the point where equilibrium is again reached atthe lower system power output.

In determining the amount of moderation provided by themoderator-coolant vapor within the reactor, the combined, weightedaverage of the vapor densities in the first, second and third passes ofthe core must be found in order to ascertain the amount of moderationbeing supplied to the reactor core at any given instant. For example, ifthe volume fractions of moderator-coolant steam spaces within thereactor is apportioned among the three passes according to the ratio ofl0.6:77.0:l2.4 in the first, second and third passes, respectively, topro vide the requisite neutron moderation and the average temperature ofeach is determined, the total amount of moderation atoms may becalculated. In the initial design of this reactor the proportion ofmoderation to be provided in the moderating pass is determined and thecorresponding temperature and pressure requirements are established, thesteam conditions in the first and third pass then being developedtherefrom.

In order to attain a negative temperature coefficient of reactivity in areactor of the type herein disclosed, it is necessary that variations inthe reactivity level of the reactor provide self-compensating feedbackcharacteristics. This feedback is provided in the reactor of the presentinvention by utilizing the very disturbances in the reactivity which itis desired to correct to vary the moderating capabilities of themoderator-coolant vapor. The more feedback so achieved, the more stablethe reactor will be. Ideally, the steam leaving the reactor outlet wouldbe used as the moderator-coolant vapor since any variation in thereactivity level of the reactor would be transmitted to the temperature,pressure conditions of the outlet steam. However, steam at the outlettemperature and pressure of this reactor is not dense enough to providesatisfactory moderation. Accordingly, some compromise of the ideal mustbe achieved in order to attain moderator-coolant densities sufiicient tomeet the neutron moderation requirements of the reactor. This isaccomplished in the reactor of the present invention by the three-passdesign described above. Accordingly, the steam entering the first passof the reactor provides a portion of the neutron moderation, asillustrated above, while at the same time responding to reactivitylevels within the reactor which would change the reactivity of the fuelelements in the first pass and thus the heat produced therein. Thisvariation in the heat generated in the first pass is transmitted to themoderator-coolant flowing therethrough and changes the characteristicsof the steam entering the second moderating pass wherein the majorportion of the reactor moderation occurs. Accordingly, if somedisturbance within the reactor causes the reactivity to increase, thepower produced in all the fuel elements, including those of the firstpass would increase and raise the temperature of the moderator-coolantvapor leaving the first pass. While this increase in temperature of themoderatorcooiant vapor in the first pass affects the reactivity of thereactor to some extent, it may not sufficiently affect the reactivity ofthe entire reactor to counteract the disurbance mentioned above.However, due to the fact that the major portion of the moderation of thereactor occurs in the moderating pass, the increased temperature of themoderator-coolant vapor entering the moderator pass is reflected in thereactivity level of the reactor and the reactor tends to return to theprior stable conditions by changing the moderator density in themoderating pass. In the above it was assumed that an increase in thetemperature and enthalpy of the steam in the reactor would produce adecrease in the density of this steam in spite of a rise in systempressures. To achieve this effect it is necessary that system piping andcomponent volumes, including the surge chamber 54 and the rate of bypassfiow through it, are in correct proportion to the reactor steam spacevolume and reactor thermal feedback. The converse would be true, ofcourse, should the reactivity of the reactor decrease with consequentchange in moderator-coolant vapor density. In this case the temperatureof the coolant would tend to drop, returning the reactivity to theprevious level. The amount of feedback in a reactor of the presentinvention is a function of the volumetric ratio of the various passes inthe reactor core. Accordingly, if 25% feedback is satisfactory formaintaining stable reactor operating conditions with the design heredisclosed, one fourth of the total number of fuel elements would bepositioned in the first pass, with the remaining three-fourths of thefuel elements being located in the third pass. Should other feedbackratios be desired they could easily be achieved by proportioning thedistribution of fuel elements between the first and third passes,keeping in mind the vapor density required in the moderating space in"order to satisfactorily moderate the chain reaction.

The surge chamber 54- in FIG. '1 whichis connected id via lines 50 and52 across orifice 48 in the pump outlet line 42 is intended tocompensate for system disturbances and thus assure reactor stability. inoperation the surge chamber 54, in combination with the orifice 48,operates as a buffer for pressure and enthalpy surges occurring in thereactor system induced by causes other than the reactor itself. Thedegree to which enthalpy surges are damped is controlled by the amountof steam passed through the surge chamber which, in turn, is determinedby the orifice size. Such a surge might occur in the reactor system ifthe system load were suddenly dropped with a resulting increase in thetemperature of the moderator-coolant vapor. While an increase intemperature would not ordinarily affect the reactivity of the reactor,inasmuch as there would generally be a corresponding rise in pressurewhich would maintain the density of the moderator-coolant vaporsubstantially con stant with a resulting constant level of reactivity, asudden rise in temperature in a portion of the reactor system remotefrom the reactor could unbalance the reactivity of the reactor. This isdue to the fact that an increase in the temperature in one portion ofthe system causes a corresponding increase in pressure, which istransmitted to all other portions of the system almost instantaneouslywhile the increase in temperature of the moderator-coolant vapor lagsbecause of inherent characteristics of the heat transfer phenomena. Thepressure in that portion of the system remote from the disturbance, iethe reactor, thus would increase without a corresponding vaportemperature rise, thereby increasing the density of the moderatoncoolantvapor and the reactivity of the reactor. However, by utilizing a surgechamber 54, a rapid change in the system pressure in one portion of thereactor system resulting from a sudden change in the vapor temperatureis damped so that the change in pressure is reduced to the point whereits effect is of little consequence. A temperature bufler chamber, whichwill be illustrated hereinbelow, is also provided in the reactor todampen any sudden changes in the moderator-coolant vapor entering thereactor, thereby further promoting further reactor stability, ifrequired.

It should be noted that in the system illustrated in FIG. 1 themoderator-coolant vapor pump 46 is, in the fluid fiow sense, situated inthe reactor system between the parallel connected superheater-reheaterheat exchangers 34 and 36 and the boiler heat exchanger 46 to provideassurance that the moderator-coolant vapor can be maintained within thesupercritical region in all portions of the system. Inasmuch as thelowest system pressure occurring in the reactor circuit occurs at thepump inlet, it would be undesirable to have, at this same point, thelowest system temperature. This would be the case if the pump werepositioned between the boiler heat exchanger 46 and the reactor inlet 34since such an arrangeiment might result in the moderator-coolant vapordropping into the subcritical temperature and pressure range, with thepossibility of condensation of the vapor and its attendant difficultiesduring large system transients. However, when greater system pressuresare permissible the pump could be located between the final heatexchanger 45 and the reactor 22. (in FIG. 1) with a material reductionin system pumping power requirements as a result of the increaseddensity of the moderator-coolant fluid at this point.

While only a closed cycle arrangement has been described herein, an opencycle of the Loetfler type may also be utilized as set forth in theabove referenced copcnding application.

A specific embodiment of a preferred arrangement of a reactor of thepresent invention is illustrated in FIG- URES 2 to 9 wherein the reactorcomprises a pressure vessel 22, which consists of an elongatedcylindrical shell portion closed at the lower end by hemispherical head68 "having "an outlet flange 7% integrally attached thereto.

The cylindrical shell 66 terminates at the upper end in a closure flange72. An upper closure head 74 having a hemispherical head portion 76 anda closure flange 78 is adapted to close the upper end of the pressurevessel with the closure flange 78 of the head mating with the closureflange 72 of the pressure vessel and being secured thereto by aplurality of circumferentially spaced studs 80. An inlet nozzle 82 isdisposed in the upper portion of the closure head 74 and is connected toinlet line 30 of the reactor system (in FIG. 1). The juncture of theclosure head 74 with the closure flange 72 is provided with apressure-tight seal 84 which may be of a semi-toroidal ring-type as iswell known in the art. The cylindrical shell 66 of the pressure vesselmay be provided with a plurality of external bands generally indicatedas at 86 in order to more economically accom modate the comparativelyhigh internal pressure, as is well known in the art.

The interior surface of the pressure vessel is provided with a layer 88of thermal shielding material such as CaH and steel or other thermalshielding material as is well known in the art. Such thermal shieldingmaterial extends through the height of the cylindrical portion 66 of thepressure vessel and extends along the inner surface of the lower head 68to the inner circumference of flange 70. Additional thermal shieldingmaterial is provided at each end of the cylindrical section as will bedisclosed later. A layer of neutron reflecting material 90 generallycoextensive with the cylindrical shell portion 66, for instancegraphite, abuts the inner surface of the thermal shielding material 88and is supported by an annular support flange 92 secured to the innersurface of the pressure vessel.

As shown in FIG. 2, the interior of the reactor is divided into threevertically superimposed portions, the inlet plenum chamber 94, thecentral core region 95 and the outlet plenum chamber 96, the upper andlower tube sheets 97 and 98, respectively, serving to form the threezones. The upper or inlet plenum chamber 94 is formed within the upperportion of the pressure vessel by the upper tube sheet 97 which issupported by a cylindrical skirt member 100 from an outwardly extendingflange member 102 supported within a ciroumscribing .groove 104 formedin the upper closure flange 72 at its juncture with the closure head 74.A semi-toroidal sealing ring 106 is provided between the flange member102 and the closure flange 72 preventing any leakage therebetween. Asecondary upper tube sheet 110 (see FIG. is supported and spaced fromthe upper tube sheet 97 by circumferentially disposed bolts 108 andspacers 109. A plurality of tubes 112 extend through both upper tubesheets 97 and 110 and through the central core region 95 of the reactorvessel. At their upper ends they are secured .to the upper tube sheet 97with a sliding fit within the secondary upper tube sheet 110.Additionally a plurality of tubes 114 terminate between the upper tubesheets 97 and 110, having a sliding fit through the secondary upper tubesheet, and extend through the core region 95 and through the lower tubesheet 98, to which they are secured. Alternatively tubes 114 may besecured to the secondary upper tube sheet 110 which is then supportedonly by bolts 108 so as to permit differential movement.

The lower tube sheet 98 cooperates with a U-shaped plate member 116 (seeFIG. 2) to form a pressureatight space designated as outlet header 118which is connected by a plurality of connecting headers 120 to an inneroutlet nozzle 122. The nozzle is positioned to extend vertically frombelow the outlet header 118 to an outlet nozzle 124. The outlet nozzle124 is provided at its upper end with an outwardly extending flangemember 126 which is adapted to seal against the inner surface of outletflange 70. The lower end of outlet nozzle 124 is then connected tooutlet line 32 (see FIG. 1). A split retaining ring 128 circumscribesthe outer circumference of outlet nozzle 124 abutting the outer face ofthe outlet flange 70, and extending into groove 130 formed in the outersurface of the outlet nozzle 124. To rigidly secure the outlet nozzleagainst movement the ring 128 is attached to the outlet flange 70 bysuitably spaced bolts (not shown). A semi-toroidal ring-type seal 131 isattached between the outlet nozzle 124 and the outlet flange 7 0 toprovide a fluid-tight seal therebetween.

As illustrated, a layer of neutron reflecting material 132 is supportedfrom the lower surface of the secondary upper tube sheet 110; a similarlayer of neutron reflecting material 134 is supported on the uppersurface of the lowor tube sheet 98. These reflecting layers inconjunction with the wall reflecting material provide the central coreregion with complete encasement by the reflector material. Thermalshielding material is also provided at the ends of the central coreregion 95 to protect the ends of of the pressure vessel from excessivethermal heating effects. The upper thermal shielding material 136 isplaced within the inlet plenum chamber 94 and comprises an annularportion spaced above and away from the upper tube sheet 97 together witha circular portion spaced above and away from the aforementioned annularportion to provide fluid flow area from the reactor inlet to the uppertube sheet. The lower thermal shielding material 138 is located in thespace between the under side of the outlet header 118 and the upper endof the inner outlet nozzle 122.

A collection chamber 140 is supported in the inlet plenum chamber 94 bya circumferential flange 141 on the upper surface of flange member 102.The collection chamber has a generally annular shape being open at thetop to the inlet plenum chamber 94 and closed at its lower extremity,the center of the annulus providing a flow passageway for the inletmoderator-coolant vapor from the inlet nozzle '82 to the upper tubesheet 97. An inlet flow baffle 142, having a double conical shape, isdisposed concentric with the center of the condensate collection chamberand the inlet nozzle 82 to direct the incoming moderator-coolant vaporflow stream so that it does not flow directly through the centralpassage. The deflector baflle and the condensate collection chamberprovide the temperature buffer chamber described above and minimize theamount of condensed moderator-coolant vapor flowing into the core regionof the reactor when the reactor system is shut down.

While not shown, means may be provided for passing coolant fluidthroughout the neutron reflecting material and the thermal shieldingmaterial during reactor operation to maintain these materials at thedesired temperature.

Assembly of the internals of this reactor comprises the following steps.Prior to being placed within the reactor the tubes 112 are secured totube sheet 97 and tubes 114 to tube sheet 98 with the tube sheetsuitably disposed. The lower tube sheet 98, the outlet header 118 andassociated outlet nozzle 122 are supported from the upper tube sheet 97by temporary suspension bolts (not shown) through the tubes 114. Duringassembly of the tubes 112 and 114 to their respective tube sheets, tubespacers 162 (later described) are installed, the upper and lowerportions of the neutron reflecting material 132 and 134 are assembledand the lower portion 138 of the thermal shielding material ispositioned. This entire assembly is then lowered into the reactor vesseland is suspended from the upper flange 72 by the circumferential flange102 which is supported in the corresponding groove 104. The uppersealing ring 106 is then installed as is the lower sealing ring 131. Theretainer ring 128 is then installed securing the outlet nozzle 124 tothe outlet flange 70. It should be noted that this arrangement providescompensation for differential thermal expansion of the various parts ofthe reactor internals since the lower portion comprising tubes 114,outlet header 118 and the outlet nozzle 124, are rigidly secured only tothe outlet flange 70. These components thus are free to expand upwardlywith the tubes 114 moving through the secondary upper tube sheet 110.The upper tube sheet 97 and associated portion of the reactor internalsare supported only at the upper extremity by flange member 102 throughthe instrumentality of groove 104- in flange 72, the tubes 112 beingfree :to expand through the secondary upper tube sheet 111 since thereis sufiicient clearance provided to permit the relative motion.Clearance is also provided at the lower end of tubes 112, adjacent tubesheet 93 to accommodate this expansion.

After installation of the reactor internals the fuel elements areinserted and the upper thermal shielding material 136, the condensatecollection chamber 140, and battle 142 are positioned and the upperclosure head 74 is attached to the pressure vessel.

In operation the moderator-coolant vapor enters the reactor throughinlet nozzle 82, flows around the inlet flow bafile 142 and through thecentral passage provided in the collection chamber 1140 around the upperthermal shielding material 136 to the inlet ends of tubes 112 in theupper tube sheet 97. The moderator-coolant vapor then flows through thetubes 112 discharging in the lowest portion of the central core region95 adjacent the neutron reflecting material 134 and then flowing upwardaround the external surface of all the tubes 112 and 114, thence back tothe space between the tube sheets 97 and 110 through appropriatepassages in tube sheet 110 and reflector 132 (not shown) and thence intotubes 114. After flowing through tubes 114- the moderator-coolant vapordischarges into the outlet header 118 and on into outlet nozzle 124 viathe connecting headers 12%) and the inner outlet nozzle 122.

The tubular fuel elements of the reactor are suspended within tubes 112and 114 with an annular space formed between the exterior surface of thefuel element and the inner surface of the respective tubes through whichmoderator-coolant vapor flows absorbing heat in direct heat transferrelationship with the fuel elements. The fuel elements within the tubes112 comprise the first pass 24, discussed above with relation to FIG. 1,while the space about all of the tubes Within the central core region 95comprises the second or moderating pass 26, while the fuel elementswithin tubes 114 form the third pass 28.

A portion of the cross-sectional view of the reactor showing a plan viewof fuel element bundles 144- may be seen in FIG. 3. The fuel elementbundles have a generally hexagonal cross section with fuel pins arrangedin a triangular pitch throughout the cross section of the reactor core.FIG. 4 shows an enlarged plan view of a single fuel element bundlehaving a central handling knob 146 and a hexagonal fuel element tubesheet 148. Referring also to FIG. 5, it may be seen that the fuelelement bundle tube sheet 148 is provided with a plurality of openingstherethrough, through which the fuel pins 160 are suspended fromenlarged'end caps 150 into tubes 112 and 114. A plurality of smallerholes 152 are also provided through the fuel element tube sheet 144 forpassage of the moderator-coolant vapor therethrough into tubes 112. Asecond fuel element bundle tube sheet 155 shown adjacent the fuelelement tube bundle sheet 143 is free to move along the fuel pins sothat upon removal of the fuel element bundle 144 from tubes 112 and 114this second tube sheet 156 will become positioned at the lower end ofthe fuel element bundle thus providing spacing for the lower ends of thefuel pins 160. A plurality of spacers 154 are provided on the lowersurface of the fuel element tube sheet 155 which space it and tube sheet148 from the upper surface of tube sheet 97 thus permitting a flow areatherebetween for the moderator-coolant vapor. It should be noted thatthe fuel pins suspended within tubes 114-, which form the third pass ofthe reactor, are provided with an enlarged portion 158 at their upperextremity to prevent excessive leakage of moderator-coolant vapor fromthe upper plenum chamber directly into the third pass tubes 114.

A cross-sectional View of a portion of the fuel elements in the centralcore region is shown in FIG. 6, wherein it is seen that the fuelelements are divided between the third pass tubes 114 and first passtubes 112 in a ratio of 3:1. It is also seen that a layer of insulation166 is positioned around the outer surface of the third pass tubes 114,limiting the heat transfer between fuel elements therein and themoderator-coolant vapor in the moderating pass 26 so that the amount ofmoderation will be determined largely by the heat pickup of themoderatorcoolant vapor in the first pass and permit only a very minorabsorption of heat by the moderator-coolant vapor in the moderatingpass. Also shown in FIG. 6 is one row of tube spacers 162 which arespaced throughout the height of the central core region 95, as seen inFIG. 2. Each spacer is comprised of two sections which are boltedtogether, as by bolts 164, around the first pass tubes 112 permittingtubes 114 to move independently therefrom. The spacers 162 are arrangedso as to form flow passages 165, having approximately the same flow areaas the annular flow passages of the first and third passes. Spacers 162.provide the tubes 112 and 114 with lateral bracing throughout theirlength while still providing for moderator-coolant vapor flowtherethrough and around all of the tubes to moderate the chain reaction.

An enlarged detail of a fuel pin within a third pass tube 114 may beseen byreferring to FIG. 7 wherein the fuel pin 160, containing aplurality of fissionable material-bearing pellets 167 as is well knownin the art is disposed within a corrugated tubular member 170 and spacedtherefrom by a plurality of spacing strips 168 extending longitudinallyalong the fuel pin and equally spaced therearound. The flow passage forthe moderatorcoolant vapor is thus the annular space between the fuelpin and the corrugated tubular member 17% A layer of thermal insulatingmaterial such as stainless steel W001 174 is disposed about the outersurface of the corrugated tube member and is encased by an outercovering 172. This thermal insulating material, as discussed above,reduces the heat transfer from the fuel pin of the third pass throughthe moderator-coolant vapor in the annular passage and the tube wall tothe surrounding moderatorcoolant vapor in the second moderating pass.While straight sided tubes may be used as shown in FIG. 5, thecorrugation of the tubular member 176 permits the use of a relativelythin wall resulting in a conservation of neutrons, while at the sametime providing satisfactory strength so that the tubular member may beselfsupporting.

The fuel pin 160 of the first pass is shown in FIG. 8 and is similar tothat of the third pass, being disposed within a corrugated tubularmember 173 which forms tubes 112;. The external surface of fuel pins 161and both the first and third passes may be provided with surfaceroughening 180 to increase the heat transfer coefficient between thefuel pin and the moderator-coolant vapor.

A modified form of third pass tube arrangement is shown in FIG. 9wherein the corrugated tubular member 17 fl is provided with insulationin the form of a plurality of layers of stainless steel foil 182 whichis encased by the outer covering 184 having a plurality of grooves ordimples 186 formed therein for rigidity.

Table I sets forth details of the reactor described above:

TABLE I Reactor output 750 mw. heat. Net plant output 321 mw. elec.U-235 loading (initial/final)* 1070.4/520.5 kg.

U-238 loading (initial/final)* 22,632/22,221 kg. Fuel enrichment(initial/final)* 4.52/2.28%.

* Based upon 95% theoretical density U02.

1 5 TABLE I-Continued Max. fuel clad surface temp.:

Number of fuel pins in 1st pass 1450. Number of fuel pins in 3rd pass4350.

Pin pitch (triangular) 1.30 in. Active pin length 12 ft. Average fueltemperature 2100 F. Graphite reflector thickness 17.5 in.

CaH thermal shield thickness 3.5 in.

1st pass assembly:

Fuel tube O.D 0.5275 in. Shroud tube centerline diameter 0.978 in.

3rd pass assembly:

Fuel tube O.D. 0.5275 in. Shroud tube centerline diameter 0.717 in.Insulation foil equivalent thickness 0.002 in. Insulation sleeve I.D.0.994 in.

Core pass steam conditions 1st 2nd 3rd pass pass pass Inlet Steam Temp,F 692 717 719 Inlet Steam enthalpy, B.t.u./lb. 762. 4 985. 2 972. 8Inlet Steam density, lb./lt. Inlet Steam velocity, ft./see Inlet SteamPress, p.s.i.a Outlet Steam Temp, F... Outlet; Steam enthalpy, B. .IlbOutlet Steam density, lb./it. Outlet Steam velocity, it./sec OutletSteam Press, p.s.i.a- 3. 400 3, 400 3, 300 Heat Absorbedmw. heat 183. 440. 0 526. 6

The superiority of the reactor of the present invention results from theadvantages of using high pressure steam for cooling, moderation andcontrol. The moderating power of supercritical steam permits developmentof compact reactor cores capable of considerably higher power output perunit of core volume within currently available pressure vessel sizes.Variation of moderator steam density during core life is advantageouslyused to achieve reactivity control thereby simplifying fabrication ofthe system. Furthermore, control rods may be replaced by a system ofpositive displacement pumps and relief valves with sensitivity amenableto close regulation, as described above, to produce controlled changesin reactivity simply by varying reactor system steam inventory anddensity. Maximum clad surface temperatures in this reactor are unusuallylow due to the absence of control rods and the flat radial power plus askewed axial power profile in which the power peaks an appropriateamount in the cold end of the core. The amount of peaking is controlledby placing nonmoderating material, such as ZrO or A1 0 in the insulatorspacebetween tubes 114 and the outer covering 172. This, in combinationwith high reactor system efliciency and excellent coolantcharacteristics, permits a high power densit as set forth above,resulting in possible throw-away fuel cycle economics.

A further advantage of the present invention resides in the ability ofthe reactor to even out variations in the reactor neutron flux and powerdensity level of the core. This occurs since any variation in radialpower distribution of the core will produce a corresponding variation inmoderator-coolant vapor density distribution since the amount of heatadded to the moderator-coolant vapor in a specific zone of the reactorwill determine the density of the moderator-coolant vapor in the samespecific zone of the moderating space and thus regulate the power of thereactor in these zones. This flattening of power density is illustratedin FIG. 12 wherin trace line 192 illustrates the power distribution in areactor having uniform moderating density and fuel enrichment throughoutthe core, while trace line 194 illustrates the relative power density ofa reactor of the present invention.

Additional advantages of the present invention reside in the fact thatthe control of this reactor follows output load demands much moreclosely than do reactors of the prior art. This is due to the fact thatupon a change in output load the returning moderator-coolant vaporentering the reactor quickly reflects the load change and immediatelyaffects the reactivity of the reactor, while the reactors of the priorart are subject to several delay factors including the lag in signaltime between sensing the change in the output load and the operation ofthe control rods. Another adverse effect common to reactors of the priorart, but which is significantly absent in a reactor of the presentinvention, is the fly-wheel effect of the coolant fluid which results ina lag in the speed of response to load change due to the time requiredto change the temperature of a large mass of coolant liquid in thereactor as compared to the sensitivity and speed of responsecharacteristic of the vapor coolant of this invention. As a result ofthis fly-wheel effect, practical considerations have made it necessaryfor the control rods of the reactors of the prior art to be regulated toovercompensate for load changes so that the reactivity of the reactortends to hunt until the final temperature and power level are achieved.Conversely, as a result of the low mass of the moderator-coolant vaporand its quick reaction to changes in output load, there is little changein the output temperature of the reactor resulting from output loadchanges.

A variation in the reactor system illustrated in FIG. 1 is shown in FIG.13 whereby wider output variation may be attained by circulating alarger quantity of moderatorcoolant vapor through the first pass of thereactor while maintaining the original flow rate in the second and thirdpasses. This is accomplished by extracting a portion ofmoderator-coolant vapor from the reactor at the end of the first passvia line 200 through valve 202 into the inlet line 38 of pump 40. Inthis way relative flow of moderator-coolant vapor through the first passis increased, decreasing the heat pickup per pound of vapor andincreasing the average density of the vapor throughout the reactor andthereby increasing reactivity and the power output. However, if by-passline 200 is used alone, the reactor outlet temperature will decreasethus increasing the density within the core for the purpose ofincreasing power and prolonging core life.

If it should be desired to provide a more efficient reactor system byextracting a greater amount of heat in the boiler 46, lowering themoderator-coolant vapor temperature at the reactor inlet, it isnecessary to increase the density of the moderator-coolant vapor in thesecond pass while maintaining the desired heat pickup Within the reactorand keeping the outlet temperature of the reactor constant and wouldordinarily result in a lower flow rate through the reactor. This causesdifiiculties in the sizing of the flow passages through the tubes aroundthe fuel elements since the mass rate of flow of the moderator-coolantvapor determines the heat transfer characteristics and any variation inthe flow rate would change the heat pickup of the vapor. However, withthe arrangement shown in FIG. 13 it is possible to lower the density ofthe moderator-coolant in the second pass and yet maintain the flow ratesin the first and third passes relatively constant while also maintainingthe reactor outlet temperature substantially constant. This isaccomplished by removing a portion of the moderatorcoolant vapor fromthe end of the first pass via by-pass line 2% through valve 202 andinlet line 38 to pump 4tl and introducing a corresponding amount ofmoderator vapor back into the reactor at the beginning of the third passthrough by-pass line 264 through valve 206 and into the reactor via line2%. Since the extracted vapor is mixed with much hotter vapor leavingheat exchangers 34 and 36, the vapor reintroduced into the reactor vialines 204 and 253 is also at a higher temperature than would be thevapor flowing normally from the first pass resulting in a substantiallyconstant reactor outlet temperature and outlet flow rate.

Conversely, should it be desirable to decrease the amount of heatextracted from the moderator-coolant vapor by heat exchangers 34, 36,and 46, while maintaining the flow rates and the reactor outlettemperature substantially constant, it would be necessary to decreasethe density of the moderator-coolant vapor in the second pass. Toaccomplish this, a portion of the moderatorcoolant vapor is extractedfrom the reactor at the end of the first pass via by-pass line 2% and isreintroduced into the first pass of the reactor vialine 204 throughvalve 210 and line 212 resulting in a somewhat higher inlet temperatureto the reactor. In order to keep the outlet temperature from rising, aportion of cooler moderator vapor is introduced at the beginning of thethird pass from the outlet of the heat exchanger 46 via bypass line 214through valve 216, thus maintaining the flow through the reactor withinsatisfactory limits and the reactor outlet temperature constant.Accordingly, the present invention may be arranged for very flexibleoperation with wide ranges of output obtainable while maintainingsatisfactory flow rates and a substantially constant outlet temperature.

Many variations present themselves when the various types ofmoderator-coolant vapors are considered, the only limitations placedupon these fluids being that they have desirable heat transfercharacteristics and that they be compatible for use with structuralmaterials available at the temperatures and pressures contemplated. Itis necessary, of course, that the moderator-coolant vapor contain asufiicient quantity of hydrogen atoms, either in a free state or incombination, to provide the necessary amount of neutron moderation atpractical working temperatures and pressures. Further, it is necessarythat the moderator-coolant vapor be in a vaporous state at the workingtemperatures and pressures in order to provide the requisite variationin fluid density. This requirement for vaporized moderator-coolant vaporis also necessary in order to make the requisite safety scram obtainablemerely by venting the reactor circuit to a lower pressure through aquick opening relief valve.

In line with the above, it may readily be seen that hydrocarbon vaporscan be used as the moderator-coolant vapor. While the operatingpressures of a hydrocarbon moderator-coolant vapor would generally belower than tlBse required using steam moderation, the basic theory andmode of operation would still be the same. Additionally, the use ofhydrocarbon vapors would make possible the adoption of neutronirradiation in the refining and production of petroleum products orother chemical compounds. For example, acetylene or benzene vapor couldbe used as the moderator-coolant vapor while undergoing changes tohigher polymer hydrocarbons by means of chemonuclear reactions.

A further modification of the present invention envisions the use ofonly fissionable material-bearing fuel elements in the reactor core.Such a reactor would be one in which neutron economy is of onlysecondary in- 18 terest, the primary objective being a reactor having ahigh power output from a relatively small and compact size. In thisadaptation the variation in the moderatordensity would change only theamount of neutron leakage out of the reactor core to variably controlthe fissiontype chain reaction.

An additional modification is realized when the method of control of thereactor herein described is combined with the prior art method ofmoderation, namely, the use of a static, built-in type of moderator,such as graphite or zirconium-hydride. Such an arrangement would besimilar to the graphite-moderated, gas-cooled reactors,

wvell known in the art, except that the moderation provided by thegraphite would not be permitted to be sufiL cient to make the reactorcore critical. The final amount of moderation necessary to make thereactor critical would be supplied by the hydrogembearingmoderatorcoolant vapor described above. In this way it would still bepossible to variably control, and even to shutdown the chain reaction byvarying the density of the moderating vapor.

Additionally, the present invention may be utilized in a reactor havinga central core region arranged to undergo fission reactions as a resultof fast neutrons generated by the fission process. Such a fast coreregion would be surrounded by a thermal blanket region which, in turn,is surrounded by a neutron reflector. While a fast core region has norequirement of neutron moderation, it is controlled by the variation ofneutron moderation within the thermal bianket region. This is due to thefact that the amount of neutrons being reflected into the fast coreregion will be determined by the number of neutrons slowed by moderationin the thermal region and entering into thermal fission reactions. Suchneutron moderation as taught by the present invention results in higherneutron efficiencies for use in such a fastthermal core than would bethe case should other prior art moderating methods be used.

While in accordance with the provisions of the statutes, the best formsand modes of operation of the invention now known have been illustrated,those skilled in the art will understand that changes may be made in theform of the apparatus without departing from the spirit of the inventioncovered by the claims, and that certain features of the invention maysometimes be used to advantage without a corresponding use of otherfeatures.

What is claimed is:

1. A heterogeneous nuclear reactor comprising a pressure vessel dividedinto a plurality of passes for the flow of a fluid therethrough, amultiplicity of fissionable material-bearing fuel elements arranged in afixed lattice as a core in a first and a third pass of said reactor,said core arranged to undergo a fission-type chain reaction, means forpassing a single phase hydrogen-bearing moderating and cooling vaporthrough a first pass in direct heat transfer contact with said fuelelements therein to cool said elements and heat said vapor to apredetermined range of moderating density, means for passing said vaporthrough a second pass about all of said fuel elements in said first andthird passes to moderate said chain reaction, means for passing saidheated vapor through said third pass in direct heat transfer contactwith said fuel ele ments therein, said first pass fuel elements beingarranged to reflect changes in reactivity of said reactor by changingthe amount of heat imparted to said vapor in contact therewith toregulate the density of said vapor entering said second pass to vary thereactivity of said reactor, and means for varying the density of saidvapor in said core to regulate the chain reaction throughout the life ofsaid fuel elements.

2. A heterogeneous nuclear reactor comprising a pressure vessel dividedinto a plurality of serially arranged passes arranged for the flow of afluid therethrough, a multiplicity of fissionable material-bearing fuelelements arranged in a fixed lattice as a core in a first and a third 19pass of said reactor, said core arranged to undergo a fission-type chainreaction, means for passing steam as a single phase hydrogen-bearingmoderating and cooling vapor through said first pass in direct heattransfer contact with said fuel elements therein to cool said elementsand heat said steam to a predetermined range of moderating density,means for passing said steam through a second pass in indirect heattransfer contact with all of said fuel elements to moderate said chainreaction, and means for passing said heated steam through said thirdpass in direct heat transfer contact with said fuel elements therein tofurther heat said steam, said first pass fuel elements being arranged toreflect changes in reactivity of saidreactor by changing the amount ofheat imparted to said steam in contact therewith to regulate the densityof said steam entering said second pass to vary the reactivity of saidreactor.

3. A heterogeneous nuclear reactor comprising a pressure vessel dividedinto a plurality of serially arranged passes arranged for the flow of afluid therethrough, a multiplicity of fissionable material-bearing fuelelements arranged in a fixed lattice as a core in a first and a thirdpass of said reactor, said core arranged to undergo a fission-type chainreaction, means for passing steam as a single phase hydrogen-bearingmoderating and cooling vapor through said first pass in direct heattransfer contact with said fuel elements therein to cool said elementsand heat said steam to a predetermined range of moderating density,means for passing said steam through a second pass in indirect heattransfer contact with all of said fuel elements to moderate said chainreaction, means for passing said heated steam through said third pass indirect heat transfer contact with said fuel elements therein to furtherheat said steam, said first pass fuel elements being arranged to reflectchanges in reactivity of said reactor by changing the amount of heatimparted to said steam in direct heat transfer contact therewith toregulate the density of said steam entering said second pass to vary thereactivity of said reactor, and means for varying the concentration ofvaporous hydrogen in said core to regulate said chain reactionthroughout the life of said fuel elements.

' 4. A method of controlling, moderating, and cooling a heterogeneousnuclear reactor having a multiplicity of fissionable material-bearingfuel elements arranged in a fixed lattice as a core to undergo afission-type chain reaction, said core having a plurality of fixedregular flow channels therethrough, comprising the steps of introducingH and D 0 bearing supercritical steam as a moderating and cooling vaporin direct heat transfer contact with one group of said fuel elements tocool said elements and heat said steam to a predetermined range ofmoderating density, then passing said steam at said moderating densityin indirect heat transfer contact with all of said fuel elements in saidcore to moderate said chain reaction then passing said heated steam indirect heat transfer contact with the remaining group of said fuelelements to be further heated, allowing the amount of moderation of saidchain reaction to be automatically regulated by the heat imparted tosaid moderating and cooling steam by said first group of fuel elements,and regulating said chain reaction throughout the life of said fuelelements by varying the density of said steam in said reactor and byvarying the proportion of H 0 and D 0 in said steam.

5. A heterogeneous nuclear reactor comprising a pressure vessel dividedinto a plurality of serially arranged passes arranged for the flow of afluid therethrough, a multiplicity of fissionable material-bearing fuelelements arranged in a fixed lattice as a core in a first and a thirdpass of said reactor, said core arranged to undergo a fission-type chainreaction, means including a closed circult adapted to pass H 0 and D 0bearing supercritical steam as a moderating and cooling vapor throughsaid first pass in direct heat transfer contact with said fuel elementstherein to cool said elements and heat said steam to a predeterminedrange of moderating density, means for passing said steam through asecond pass in indirect heat transfer contact with all of said fuelelements to moderate said chain reaction, means for passing said heatedsteam through said third pass in direct heat transfer contact with saidfuel elements therein to further heat said steam, said first pass fuelelements arranged to reflect changes in reactivity of said reactor bychanging the amount of heat imparted to said steam in direct heattransfer contact therewith to regulate the densiy of said steam enteringsaid second pass to vary the reactivity of said reactor, means forvarying the proportion of H 0 and D 0 in said steam and for varying thedensity of said earn in said core to regulate said chain reactionthroughout the life of said fuel elements.

6. A method of controlling, moderating, and cooling a heterogeneousnuclearreac tor having a multiplicity of fissionable material-bearingfuel elements arranged in a fixed lattice therein as a core to undergo afission-type chain reaction, said core being arranged to have a positivecoolant density coeflicient of reactivity, said core having a pluralityof fixed regular flow channels therethrough, said flow channels beingarranged in a plurality of serially arranged passes for the flow of amoderator coolant vapor therethrough,said fuel elements being arrangedin at least one of said passes, comprising the steps of circulating asingle phase hydrogen-bearing moderating and cooling vapor through saidreactor while maintaining said vapor in a single phase throughout saidcore, passing said moderating and cooling vapor through a first passover said fuel elements therein to cool said elements and heat saidvapor to a predetermined range of moderating density, predominatelymoderating said chain reaction by passing said vapor at said moderatingdensity through a second pass, and allowing the amount of moderation ofsaid chain reaction to be automatically regulated by the heat impartedto said vapor by said fuel elements in said first pass.

7. The method of operating a heterogeneous nuclear reactor according toclaim 6 wherein the chain reaction is regulated throughout the life ofsaid fuel elements by controllably varying the density of said vapor insaid reactor to vary the concentration of hydrogen in said core.

8. The method of operating a heterogeneous nuclear reactor according toclaim 6 wherein the chain reaction is regulated throughout the life ofsaid fuel elements by controllably varying the concentration of hydrogenin said vapor.

9. The method of operating a heterogeneous nuclear reactor according toclaim 6 wherein the vapor in the first pass is in direct heat transfercontact with the fuel elements therein and the vapor in the second passis in indirect heat transfer contact with the fuel elements.

10. A heterogeneous nuclear reactor comprising a pressure vessel dividedinto a plurality of serially arranged passes arranged for the flow of afluid therethrough, a multiplicity of fissionable material-bearing fuelelements arranged in a fixed lattice as a core in a first pass of saidreactor, said core being arranged to undergo a fission-type chainreaction, means for passing a single phase hydrogen-bearing moderatingand cooling vapor through said first pass in direct heat transfercontact with said fuel elements therein to cool said elements and heatsaid vapor to a predetermined range of moderating density, and means forpassing said vapor through a second pass in indirect heat transfercontact with said fuel elements to moderate said chain reaction, saidfirst pass fuel elements being arranged to reflect changes in reactivityof said reactor by changing the amount of heat im-' parted to said vaporin contact therewith to regulate the density of said vapor entering saidsecond pass to vary the reactivity of said reactor.

11. A method of controlling, moderating, and cooling a heterogeneousnuclear reactor having a multiplicity of fissionable material-bearingfuel elements arranged in a fixed lattice as a core to undergo afission-type chain reaction, said core being arranged to have a positivecoolant density coefficient of reactivity, said core having a pluralityof fixed regular flow channels therethrough, comprising the steps ofintroducing single phase steam in direct heat transfer contact With onegroup of said fuel elements to cool said elements and heat said steam toa predetermined range of moderating density, then passing said steam atsaid moderating density in indirect heat 1 transfer contact with all ofsaid fuel elements in said core to moderate said chain reaction, thenpassing said heated steam in direct heat transfer contact with theremaining group of said fuel elements to be further heated, allowing theamount of moderation of said chain reaction to be automaticallyregulated by the heat imparted to the moderating and cooling steam bythe first group of said fuel elements, and regulating said chainreaction throughout the life of said fuel elements by varying the con-22 centration of hydrogen in said steam and by varying the density ofsaid steam in said reactor.

12. The method of operating a heterogeneous nuclear reactor according toclaim 11 wherein said steam is in the supercritical state.

References Cited by the Examiner FOREIGN PATENTS 1,273,431 9/1961France. "1,079,753 4/1960 Germany.

749,064 5/ 1956 Great Britain.

OTHER REFERENCES HW-59684, Supercritical Pressure Power Reactor, Mar.18, 1959, pages 1-37.

CARL D. QUARFORTH, Primary Examiner.

REUBEN EPSTEIN, Examiner.

1. A HETEROGENEOUS NUCLEAR REACTOR COMPRISING A PRESSURE VESSEL DIVIDEDINTO A PLURALITY OF PASSES FOR THE FLOW OF A FLUID THERETHROUGH, AMULTIPLICITY OF FISSIONABLE MATERIAL-BEARING FUEL ELEMENTS ARRANGED N AFIXED LATTICE AS A CORE IN A FIRST AND A THIRD PASS OF SAID REACTOR,SAID CORE ARRANGED TO UNDERGO A FISSION-TYPE CHAIN REACTION, MEANS FORPASSING A SINGLE PHASE HYDROGEN-BEARING MODERATING AND COOLING VAPORTHROUGH A FIRST PASS IN DIRECT HEAT TRANSFER CONTACT WITH SAID FUELELEMENTS THEREIN TO COOL SAID ELEMENTS AND HEAT SAID VAPOR TO APREDETERMINED RANGE OF MODERATING DENSITY, MEANS FOR PASSING SAID VAPORTHROUGH A SECOND PASS ABOUT ALL OF SAID FUEL ELEMENTS IN SAID FIRST ANDTHIRD PASSES TO MODERATE SAID CHAIN REACTION, MEANS FOR PASSING SAIDHEATED VAPOR THROUGH SAID THIRD PASS IN DIRECT HEAT TRANSFER CONTACTWITH SAID FUEL ELEMENTS THEREIN, SAID FIRST PASS FUEL ELEMENTS BEINGARRANGED TO REFLECT CHANGES IN REACTIVITY OF SAID REACTOR BY CHANGINGTHE AMOUNT OF HEAT IMPARTED TO SAID VAPOR IN CONTACT THEREWITH TOREGULATE THE DENSITY OF SAID VAPOR ENTERING SAID SECOND PASS T VARY THEREACTIVITY OF SAID REACTOR, AND MEANS FOR VARYING THE DENSITY OF SAIDVAPOR IN SAID CORE TO REGULATE THE CHAIN REACTION THROUGHOUT THE LIFE OFSAID FUEL ELEMENTS.